The use of radioactive materials in diagnostic medicine has been readily accepted because these procedures are safe, minimally invasive, cost effective, and they provide unique structural and/or functional information that is otherwise unavailable to the clinician. The utility of nuclear medicine is reflected by the more than 13 million diagnostic procedures that are performed each year in the U.S. alone, which translates to approximately one of every four admitted hospital patients receiving a nuclear medical procedure. [See, Adelstein et al. Eds., Isotopes for Medicine and the Life Sciences; National Academy Press, Washington, D.C. (1995); Wagner et al., “Expert Panel: Forecast Future Demand for Medical Isotopes,” Department of Energy, Office of Nuclear Energy, Science, and Technology (1999); Bond et al., Ind. Eng. Chem. Res. (2000) 39:3130–3134.] More than 90 percent of these procedures are for diagnostic imaging purposes and use technetium-99m (99mTc) as the radionuclide. 99mTc possesses a unique combination of convenient production and availability, coupled with appropriate nuclear decay mode, decay energy, and chemical reactivity. These properties enable 99mTc to be coupled to biolocalization agents that permit the imaging of many diseases and virtually every part of the human anatomy. [See, Bremer, Radiochim. Acta (1987) 41:73–81; Steigman et al., The Chemistry of Technetium in Medicine, National Academy Press: Washington, D.C., (1992); Schwochau, Angew. Chem. Int. Ed. Eng. (1994) 33:2258–2267.]
The typical life cycle of a medical radionuclide, such as 99mTc, commencing with raw material acquisition and proceeding through nucleogenesis of a radiochemical and clinical administration of the purified and sterile radiopharmaceutical is depicted schematically in FIG. 1. Technetium-99m is used as a specific example in this discussion because the vast majority of all nuclear medical procedures utilize this radionuclide, and aspects of new production technologies are typically compared to this successful model. The 99mTc desired “daughter” is formed by β1− (or negatron) decay of the molybdenum-99 (99Mo) “parent”, which forms as a result of the fission of uranium-235 in a nuclear reactor. [See, Bremer, Radiochim. Acta (1987) 41:73–81; Schwochau, Angew. Chem. Int. Ed. Eng. (1994) 33:2258–2267; Boyd, Radiochim. Acta (1987) 41:59–63; and Ali et al., Radiochim. Acta (1987) 41:65–72.]
Molybdenum-99 is separated from its nucleosynthesis precursors and byproducts during “Chemical Processing”, which represents the last stage as a “Radiochemical” according to FIG. 1. Such “Radiochemicals” encounter far less stringent regulation of the chemical and radionuclidic purity and no biological requirements (e.g., sterility and nonpyrogenicity) are enforced. Upon completion of “Chemical Processing”, which includes generator fabrication, the 99Mo/99mTc pair has become a “Radiopharmaceutical” (according to FIG. 1) and is now subject to rigorous control of the chemical purity, radionuclidic purity, sterility, and nonpyrogenicity.
Chemical purity is vital to a safe and efficient medical procedure, because the radionuclide is generally conjugated to a biolocalization agent prior to use. This conjugation reaction relies on the principles of coordination chemistry wherein a radionuclide is chelated to a ligand that is covalently attached to the biolocalization agent. In a chemically impure sample, the presence of ionic impurities can interfere with this conjugation reaction. If sufficient 99mTc, for example, is not coupled to a given biolocalization agent, poorly defined images are obtained due to insufficient photon density localized at the target site and/or from an elevated in vivo background due to a specific distribution in the blood pool or surrounding tissues.
Regulation of radionuclidic purity stems from the hazards associated with the introduction of long-lived or high energy radioactive impurities into a patient, especially if the biolocalization and body clearance characteristics of the radioactive impurities are unknown. Radionuclidic impurities pose the greatest threat to patient welfare, and such interferents are the primary focus of clinical quality control measures that attempt to prevent the administration of harmful and potentially fatal doses of radiation to the patient.
In addition to the controls placed on the chemical and radionuclidic purity of a “Radiopharmaceutical”, FIG. 1 also indicates that biological requirements are instituted. The internal administration of radiopharmaceuticals obviously mandates that the pharmaceutical be sterile and nonpyrogenic, and such requirements are familiar to medical practitioners.
Complementing the favorable nuclear and chemical characteristics of 99mTc are favorable economics and the convenience with which this radionuclide can be produced to meet radiopharmaceutical specifications. Taken together, these factors have been vital to the success of nuclear medicine.
The chemistry underlying the separation of 99mTc from 99Mo relies on the high affinity of alumina (Al2O3) for molybdate-99 (99MoO42−) and its negligible affinity for pertechnetate-99m (99mTcO41−) in physiological saline solution. FIG. 2 shows a conventional 99mTc generator or “99mTc cow”, in which the 99MoO42− parent is immobilized on an Al2O3 sorbent from which the 99mTcO41− can be conveniently separated by ascending elution of a physiological saline solution into a vacuum container. [See, Bremer, Radiochim. Acta (1987) 41:73–81; Schwochau, Angew. Chem. Int. Ed. Eng. (1994) 33:2258–2267; Boyd, Radiochim. Acta (1982) 30:123–145; and Molinski, Int. J. Appl. Radiat. Isot. (1982) 33:811–819.]
The above “conventional generator” affords 99mTcO41− of adequate chemical and radionuclidic purity for use in patients and has the benefits of ease of use, compact size, and the safety of having the principal radiologic hazard (i.e., 99MoO42−) immobilized on a solid Al2O3 support. The latter benefit eases restrictions on transport of the generator to the nuclear pharmacy and simplifies manual processing by the nuclear medicine technician.
Given the preeminent position of 99mTc in nuclear medicine and the simple and effective operation of the conventional 99mTc generator shown in FIG. 2, the logic and design of this radionuclide generator have become the industry standard for nuclear medicine. This generator methodology is not, however, universally acceptable for all radionuclides, especially for those having low specific activity parent sources or those radionuclides proposed for use in therapeutic nuclear medicine. The difficulties of using the conventional generator technology with low specific activity parent radionuclides; that is, the microquantities of the parent radioisotope present as a mixture with macroquantities of the nonradioactive parent isotope(s), derive from the need to distribute macroquantities of parent isotopes over a large volume of support so as not to exceed the sorbent capacity. Large chromatographic columns are not practical for nuclear medical applications as the desired daughter radionuclide is recovered in a large volume of eluate and, as such, is not suitable for clinical use without secondary concentration. Radionuclides useful in therapeutic nuclear medicine represent unique challenges to the conventional generator technology and warrant further discussion.
The use of radiation in disease treatment has long been practiced, with the mainstay external beam radiation therapy now giving way to more targeted delivery mechanisms. By example, sealed-source implants containing palladium-103 or iodine-125 are used in the brachytherapeutic treatment of prostate cancer; samarium-153 or rhenium-188 conjugated to diphosphonate-based biolocalization agents concentrate at metastasis in the palliative treatment of bone cancer pain; and radioimmunotherapy (RIT) employs radionuclide conjugation to peptides, proteins, or antibodies that selectively concentrate at the disease site whereby radioactive decay imparts cytotoxic effects. Radioimmunotherapy represents the most selective means of delivering a cytotoxic dose of radiation to diseased cells while sparing healthy tissue. [See, Whitlock, Ind. Eng. Chem. Res. (2000), 39:3135–3139; Hassfjell et al., Chem. Rev. (2001) 101:2019–2036; Imam, J. Radiation Oncology Biol. Phys. (2001) 51:271–278; and McDevitt et al., Science (2001) 294:1537–1540.] In addition, the recent explosion of information about disease genesis and function arising from the human genome project is expected to propel RIT into a leading treatment for micrometastatic carcinoma (e.g., lymphomas and leukemias) and small- to medium-sized tumors.
Candidate radionuclides for RIT typically have radioactive half-lives in the range of 30 minutes to several days, coordination chemistry that permits attachment to biolocalization agents, and a comparatively high linear energy transfer (LET). The LET is defined as the energy deposited in matter per unit pathlength of a charged particle, [see, Choppin et al., J. Nuclear Chemistry: Theory and Applications; Pergamon Press: Oxford, 1980] and the LET of α-particles is substantially greater than β-particles.
By example, α-particles having a mean energy in the 5–9 MeV range typically expend their energy within about 50–90 μm in tissue, which corresponds to several cell diameters. The lower LET β1−-particles having energies of about 0.5–2.5 MeV may travel up to 10,000 μm in tissue, and the low LET of these β1−-emissions requires as many as 100,000 decays at the cell surface to afford a 99.99 percent cell-kill probability. For a single α-particle at the cellular surface, however, the considerably higher LET provides a 20–40% probability of inducing cell death as the lone α-particle traverses the nucleus. [See, Hassfjell et al., Chem. Rev. (2001) 101:2019–2036.]
Unfortunately, the LET that makes α- and β1−-emitting nuclides potent cytotoxic agents for cancer therapy also introduces many unique challenges into the production and purification of these radionuclides for use in medical applications. Foremost among these challenges is the radiolytic degradation of the support material that occurs when the conventional generator methodology of FIG. 2 is used with high LET radionuclides. [See, Hassfjell et al., Chem. Rev. (2001) 101:2019–2036; Gansow et al., In Radionuclide Generators: New Systems for Nuclear Medicine Applications; Knapp et al. Eds., American Chemical Society: Washington, D.C. (1984) pp 215–227; Knapp, et al. Eds., Radionuclide Generators: New Systems for Nuclear Medicine Applications American Chemical Society: Washington, D.C. (1984) Vol. 241; Dietz et al., Appl. Radiat. Isot. (1992) 43:1093–1101; Mirzadeh et al., J. Radioanal. Nucl. Chem. (1996) 203:471–488; Lambrecht et al., Radiochim. Acta (1997) 77:103–123; and Wu et al., Radiochim. Acta (1997) 79:141–144.]
Radiolytic degradation of the generator support material can result in: (a) diminished chemical purity (e.g., radiolysis products from the support matrix can contaminate the daughter solution); (b) compromised radionuclidic purity (e.g., the support material can release parent radionuclides to the eluate: termed “breakthrough”); (c) diminished yields of daughter radionuclides (e.g., α-recoil can force the parent radionuclides into stagnant regions of the support making their decay products less accessible to the stripping eluent); (d) decreases in column flow rates (e.g., fragmentation of the support matrix creates particulates that increase the pressure drop across the column); and (e) erratic performance (e.g., variability in product purity, nonreproducible yields, fluctuating flow rates, etc.).
Medical radionuclide generators typically employ three fundamental classes of sorbents for use in the conventional methodology depicted in FIG. 2: (a) organic sorbents (e.g., polystyrene-divinylbenzene copolymer-based ion-exchange resins, polyacrylate supports for extraction chromatography, and the like), (b) inorganic sorbents (e.g., Al2O3, inorganic gels, and the like) and (c) hybrid sorbents (e.g., inorganic frameworks containing surface-grafted organic chelating or ion-exchange functionalities, silica supports used in extraction chromatography, and the like).
A variety of organic sorbents, most notably the conventional cation- and anion-exchange resins, have been proposed for use in nuclear medicine generators [see, Molinski et al., Int. J. Appl. Radiat. Isot. (1982) 33:811–819; Gansow et al., in Radionuclide Generators: New Systems for Nuclear Medicine Applications, Knapp et al. Eds., American Chemical Society, Washington, D.C. (1984) pp 215–227; Mirzadeh et al., J. Radioanal. Nucl. Chem. (1996) 203:471–488; and Lambrecht et al., Radiochim. Acta (1997) 77:103–123] due to the well documented chemical selectivity [see, Diamond et al., In Ion Exchange, Marinsky Ed., Marcel Dekker, New York (1966) Vol. 1, p 277; and Massart, “Nuclear Science Series, Radiochemical Techniques: Cation-Exchange Techniques in Radiochemistry,” NAS-NS 3113; National Academy of Sciences (1971)] and the widespread availability of these materials. Unfortunately, organic-based ion-exchange resins frequently fail or are severely limited in applications using the conventional generator logic, and typically do so at radiation levels far below those needed for routine human use.
By example, polystyrene-divinylbenzene copolymer-based cation-exchange resins are used in a generator for the α-emitter 212Bi, but such materials are limited to approximately two week “duty cycles” (i.e., the useful generator lifetime accounting for chemical and physical degradation) for 10–20 mCi generators. Radiolytic degradation of the chromatographic support reportedly leads to diminished flow rates, reduced 212Bi yields, and breakthrough of the radium-224 (224Ra) parent. [See, Mirzadeh et al., J. Radioanal. Nucl. Chem. (1996) 203:471–488.] Similarly, a 213Bi generator employing an organic cation-exchange resin was limited to a shelf life of approximately one week at an activity level of 2–3 mCi of the α-emitting 225Ac parent. [See, Mirzadeh et al., J. Radioanal. Nucl. Chem. (1996) 203:471–488; and Lambrecht et al., Radiochim. Acta (1997) 77:103–123.]
With the US Food and Drug Administration's recent approval of yttrium-90 (90Y)-based RIT for widespread human use, more efficient generator technologies for this radionuclide continue to emerge. Yttrium-90 forms by β1− decay of the strontium-90 (90Sr) parent radionuclide and, thus, represents a two component separation involving Sr(II) and Y(III) (presuming a chemically pure 90Sr stock). Although a variety of 90Y production methods have been proposed, [see, Dietz et al., Appl. Radiat. Isot. (1992) 43:1093–1101; Horwitz et al., U.S. Pat. No. 5,368,736 (1994); and Ehrhardt et al., U.S. Pat. No. 5,154,897 (1992)] each technology is challenged by scale-up to Curie levels of production due to problems arising from radiolysis of the solution medium and the support matrix. The inadequacies of the solvent extraction and ion exchange-based generators for 90Y have been briefly reviewed in works proposing macrocyclic host/guest chemistry as the basis for the separation of 90Y from 90Sr. [See, Dietz et al., Appl. Radiat. Isot. (1992) 43:1093–1101; and Ehrhardt et al., U.S. Pat. No. 5,154,897 (1992).]
In these reports, the 90Sr was separated from 90Y in 3 M HNO3 on a Sr(II) selective chromatographic support containing a lipophilic crown ether. This extraction chromatographic material showed exceptional stability to γ radiation from a 60Co source, although some diminution of Sr(II) retention was noted. Unfortunately, the presence of radiolysis-induced gas pockets adversely affects the chromatographic performance of this conventional generator. Consequently, the 90Sr was stripped after each processing run to minimize radiolytic degradation of the support; however, it became increasingly difficult to achieve efficient stripping of 90Sr upon repeated use.
The use of inorganic materials in radionuclide generators has been greatly influenced by the Al2O3-based conventional 99mTc generator technology. [See, Bremer, Radiochim. Acta (1987) 41:73–81; Schwochau, Angew. Chem. Int. Ed. Eng. (1994) 33:2258–2267; Boyd, Radiochim. Acta (1987) 41:59–63; Boyd, Radiochim. Acta (1982) 30:123–145; Molinski, Int. J. Appl. Radiat. Isot. (1982) 33:811–819; Benjamins et al., U.S. Pat. No. 3,785,990 (1974); Panek-Finda et al., U.S. Pat. No. 3,970,583 (1976); Matthews et al., U.S. Pat. No. 4,206,358 (1980); Benjamins et al., U.S. Pat. No. 4,387,303 (1983); Weisner et al., U.S. Pat. No. 4,472,299 (1984); Monze et al., Radiochim. Acta (1987) 41:97–101; Forrest, U.S. Pat. No. 4,783,305 (1988); Quint et al., U.S. Pat. No. 4,833,329 (1989); Vanderheyden et al., U.S. Pat. No. 4,990,787 (1991); Evers et al., U.S. Pat. No. 5,109,160 (1992); Ehrhardt et al., U.S. Pat. No. 5,382,388 (1995); and Knapp et al., U.S. Pat. No. 5,729,821 (1998).] Although the inorganic sorbents represent an improvement with respect to radiolytic stability, such inorganic materials frequently exhibit poor ion selectivity, slow partitioning kinetics, and poorly defined morphologies that inhibit good chromatographic performance.
Using the 99mTc generator example, a two component separation (i.e., 99mTcO41− from 99MoO42− in physiological saline solution) is required, for which Al2O3 is well suited. For more complicated parent daughter relationships, however, several very different chemical species can appear between the parent and daughter in a given decay chain (e.g., a gas, a tetravalent cation, and a divalent cation separate 224Ra and 212Bi) and identifying a single inorganic sorbent capable of retaining all but the desired daughter radionuclide is difficult.
Rhenium-188 (188Re) is receiving attention as a therapeutic nuclide for the prevention of restenosis after angioplasty, for pain palliation of bone cancer, and in certain RIT procedures given the similarity of its coordination chemistry with that of its widely studied lighter congener Tc. Rhenium-188 is formed by β1− decay of tungsten-188 (188W), which is produced by double neutron capture of enriched 186W in a high flux nuclear reactor. Inefficiencies arising in the nucleosynthesis of 188W result in a low specific activity parent; that is, trace 188W is present in macroquantities of the 186W isotope. Such a mass of tungstate (WO42−) requires a large column so that the capacity of Al2O3 for WO42− is not exceeded. Large chromatographic columns yield the 188Re daughter in large volumes of solution, and a variety of secondary concentration procedures have been devised to address this shortcoming. [See, Knapp et al. Eds., Radionuclide Generators: New Systems for Nuclear Medicine Applications, American Chemical Society: Washington, D.C. (1984) Vol. 241; Mirzadeh et al., J. Radioanal. Nucl. Chem. (1996) 203:471–488; Lambrecht, et al., Radiochim. Acta (1997) 77:103–123; Knapp et al., U.S. Pat. No. 5,729,821 (1998); Knapp et al., U.S. Pat. No. 5,186,913 (1993); and Knapp et al., U.S. Pat. No. 5,275,802 (1994).]
Another seldom discussed shortcoming of the conventional generator methodology as applied to 188Re arises after the generator has concluded its duty cycle and the isotopically enriched 186W must be extracted from the bulk Al2O3 matrix. Recovery of the isotopically enriched 186W for further neutron irradiation is an important part of the economical production and use of 188Re, but the distribution of macroquantities of isotopically enriched 186W target materials over a large volume of Al2O3 inhibits cost effective processing.
The 188Re “gel generator” attempts to overcome some of the challenges faced by the inorganic Al2O3-based 188Re generator, and is based on the formation of a highly insoluble zirconyl tungstate [ZrO(WO4)] gel. [See, Ehrhardt et al., U.S. Pat. No. 5,382,388 (1995) and Ehrhardt et al., U.S. Pat. No. 4,859,431 (1989).] This concept has several advantages over Al2O3-based generators, but still suffers from the fundamental drawbacks of applying the conventional generator methodology to therapeutic radionuclides.
Although the ZrO(WO4) gel generator for 188Re can permit the use of smaller column volumes than the Al2O3-based generators, the recovery of valuable isotopically enriched 186W for subsequent irradiation is still complicated. Additional considerations include variable chromatographic behavior and flow rates, as the precipitated ZrO (WO4) solids are not of well defined particle sizes or morphologies.
The inorganic materials discussed here are not immune to radiolytic degradation, especially with the high LET radionuclides. Several early versions of the α-emitting 212Bi generator [see, Gansow et al., in Radionuclide Generators: New Systems for Nuclear Medicine Applications; Knapp et al. Eds., American Chemical Society: Washington, D.C. (1984) pp 215–227; and Mirzadeh, S. Generator-Produced Alpha-Emitters. Appl. Radiat. Isot. (1998) 49:345–349] used inorganic titanates to retain the long-lived thorium-228 parent, from which the 224Ra daughter elutes and is subsequently sorbed onto a conventional cation-exchange resin. Over time, the titanate column material succumbed to radiolytic degradation, creating fine particulates that forced separations to be performed at elevated pressures.
The hybrid sorbents can be subdivided into extraction chromatographic materials and engineered inorganic ion-exchange materials. Most of the published applications of hybrid materials have used well-known extraction chromatographic methods [see, Dietz et al., in Metal Ion Separation and Preconcentration: Progress and Opportunities; Bond et al. Eds., American Chemical Society, Washington, D.C. (1999) Vol. 716, pp 234–250], whereas the preparation and use of engineered inorganic materials is a more recent phenomenon. Extraction chromatography overcomes the poor ion selectivity and slow partitioning kinetics of inorganic materials by using solvent extraction reagents physisorbed to an inert chromatographic substrate. [See, Dietz et al., in Metal Ion Separation and Preconcentration: Progress and Opportunities; Bond et al. Eds., American Chemical Society, Washington, D.C. (1999) Vol. 716, pp 234–250.]
The radiolytic stability of extraction chromatographic supports is improved when the inert substrate is an amorphous inorganic material such as silica, with the most profound results reflected as sustainable flow rates over the generator duty cycle. Such “improved” radiolytic stability is deceptive, however, as the fundamental chemical reactions underlying the parent/daughter separation still involve molecules constructed from an organic framework that remains susceptible to radiolytic degradation. Likewise, organic-based chelating moieties have been introduced into engineered inorganic ion-exchange materials to improve ion selectivity, but such functionalities continue to suffer the effects of radiolysis.
Preliminary reports using hybrid sorbents as conventional generator supports in the production of 213Bi have appeared. [See, Lambrecht et al., Radiochim. Acta (1997) 77:103–123; Wu et al., Radiochim. Acta (1997) 79:141–144; and Horwitz et al., U.S. Pat. No. 5,854,968 (1998).] Initial investigations have relied on sorption of 225Ra by organic cation-exchange resins, which showed substantial degradation over a short period of time giving reduced yields of 213Bi, poor radionuclidic purity, and unacceptably slow column flow rates. [See, Mirzadeh et al., J. Radioanal. Nucl. Chem. (1996) 203:471–488; and Lambrecht, et al., Radiochim. Acta (1997) 77:103–123.] Initial improvements centered on sorption of the 225Ac parent of 213Bi on Dipex® Resin, an inert silica gel-based support to which a chelating diphosphonic acid diester is physisorbed. [Horwitz et al., React. Funct. Polymers (1997) 33:25–36.] The silica substrate exhibits greater radiolytic stability than the previously employed organic cation-exchange resins; however, radiolytic damage (i.e., discoloration) was observed surrounding the narrow chromatographic band in which the 225Ac parent is loaded, ultimately leading to breakthrough of the 225Ac parent. [See, Lambrecht et al., Radiochim. Acta (1997) 77:103–123; and Wu et al., Radiochim. Acta (1997) 79:141–144.]
An incremental improvement in this generator centered on reducing the radiation density by dispersing the 225Ac parent radioactivity over a larger volume of the chromatographic support, which is achieved by loading the Dipex® Resin with 225Ac in a batch mode rather than in a narrow chromatographic band. [See, Wu et al., Radiochim. Acta (1997) 79:141–144.] Unfortunately, this batch loading process is awkward and the Dipex® Resin still suffers from radiolytic degradation of the chelating diphosphonic acid diester upon which the separation efficiency relies.
Despite industry preferences for the conventional generator depicted in FIG. 2, the fundamental limitations discussed above are compounded by radiolytic degradation of the support medium when using high levels of the high LET radioactivity useful in therapeutic nuclear medicine. The severity of these limitations coupled with the ultimate liability of compromised patient safety argue for the development of alternative generator technologies, especially for therapeutically useful radionuclides.
An ideal generator technology should provide operational simplicity and convenience as well as reliable production of the theoretical yield of the desired daughter radionuclide having high chemical and radionuclidic purity. As deployed for diagnostic radionuclides, the conventional generator technology generally meets these criteria, although purity and yield have been observed to fluctuate. [See, Boyd, Radiochim. Acta (1982) 30:123–145; and Molinski, Int. J. Appl. Radiat. Isot. (1982) 33:811–819.]
The conventional generator is poorly suited, however, to systems involving low specific activity parents (e.g., the 188W/188Re generator discussed above) as well as with the high LET radionuclides useful in therapeutic nuclear medicine. In order to safely and reliably produce therapeutically useful radionuclides of high chemical and radionuclidic purity, a new paradigm in radionuclide generator technology is required. A shift in the fundamental principles governing generator technologies for nuclear medicine, and for therapeutic nuclides specifically, is supported by the fact that the inadvertent administration of the long-lived parents of high LET therapeutic radionuclides would compromise the patient's already fragile health; potentially resulting in death. Because the conventional generator strategy depicted in FIG. 2 relies on long-term storage of the parent radionuclide on a solid support that is constantly subjected to high LET radiation, no assurances can be made regarding the chemical and radionuclidic purity of the daughter radionuclide over an approximate 14–60 day generator duty cycle.
Additional support for fundamental changes in radionuclide generator technology derives from the rapidly increasing trend towards automation of routine tasks such as synthesis operations in biotechnology and high throughput blood screening in the clinical laboratory. Radionuclide generator technologies, as practiced in the nuclear pharmacies, presently lag behind in the automation of routine activities. In the nuclear medicine arena, increasing federal regulations safeguarding patient health and business competition/profitability are likely to drive the industry towards automation. The introduction of computer-controlled liquid delivery systems into the nuclear pharmacy will permit a departure from the vacuum container-based generators of FIG. 2. A reduction in the number of manual operations also serves to minimize the radiation dose to the nuclear medicine technician, while simultaneously reducing the liabilities attributable to human error.
The adverse effects of radiolytic degradation described above pose enormous challenges in the development of new therapeutic radionuclide generators. Any damage to the support material of a conventional generator compromises the separation efficiency, potentially resulting in breakthrough of the parent radionuclides and to a potentially fatal dose of radiation if administered to the patient. Such a catastrophic event is theoretically prevented by the quality control measures integrated into nuclear pharmacy operations, but any lack of safe, predictable generator behavior represents a major liability to the nuclear pharmacy, hospital, and their respective shareholders. The invention described hereinafter provides an alternative radionuclide generator technology that is capable of reliably producing near theoretical yields of medically useful radionuclides of high chemical and radionuclidic purity.